Burnup Performance Research Articles

Preliminary study of a small high-temperature gas-cooled reactor (HTGR) concept with MgO–BeO moderators

Graphite plays a crucial role as a moderator and reflector in high-temperature gas-cooled reactors (HTGRs). However, under high-irradiation conditions, graphite exhibits microcracking within the operational period. Studies have been investigated the potential of composite-based materials for replacing graphite in HTGRs. This study focused on a magnesium oxide (MgO)-based composite material as the host matrix and a homogeneously distributed beryllium oxide (BeO) as the entrained moderating phase and investigated the feasibility of the MgO–BeO as the new moderator in a small HTGR to achieve high burnup performance. In this study, the conceptual design of the HTR50S was selected as one of the candidates for the small HTGR concept. Burnup calculations and safety evaluation in HTR50S design were performed. The Monte Carlo MVP 3.0 and MVP-BURN codes were used in this study for neutronic calculations. Results revealed that a high burnup of 80 GWd/t can be achieved using a fuel composition of 6 kg heavy metal per fuel block with 17 wt% of 235U enrichment. Furthermore, a negative temperature coefficient of reactivity was achieved during the operation period.

Application of data assimilation in searching better lattice-physics parameters of fuel assembly

This study aims to apply the data assimilation technique to adjust the fuel enrichment to find a better performance of the fuel assembly lattice-physics parameters. In this study, the STARU data assimilation, which can deal with the highly non-linear system behaviors and many parameters, was implemented; and MCNP6 was used to evaluate the lattice-physics parameters of the fuel assembly. Furthermore, restriction and non-restriction sampling algorithms were employed to efficiently find the enhanced candidates for the NuScale fuel assembly. We found that the improvements were up to 0.1 w/o, and the fuel depletion of the optimal candidate was compared well with the equivalenced enrichment of the typical fuel assembly. Eventually, the conceptual design configuration was proposed using the alpha and theta requirements for adjusting the fuel rod dimensions and enrichments, methodically. Consequently, the proposed conceptual design illustrated an excellent performance of the lattice-physics parameters such as high k∞-value and low PPPF. Nevertheless, the burn-up performance of the proposed conceptual design was slightly poor compared with the NFAC-01. Future studies should examine the application of neutron absorber materials to overcome this limitation.

Scoping Studies for a Lead-Lithium-Cooled, Minor-Actinide-Burning, Fission-Fusion Hybrid Reactor Design

A feasibility study of a subcritical fission-fusion hybrid reactor using lead-lithium eutectic as a coolant and minor actinides (MAs) as fuel is presented. Such a reactor could support the fission community by transmuting MAs and the fusion community by breeding tritium. The feasibility of such a reactor for the burnup of MAs is assessed in terms of burnup performance, tritium breeding, and safety characteristics. Tandem mirrors are a promising neutron source technology, and a deuterium-tritium tandem mirror is considered here for the neutron source with power Psource = 1.13 MW assumed for scoping purposes. Subcritical reactivities from keff = 0.9800 to keff = 0.9950 were considered, representing the initial reference for subcritical reactivity and the assessed upper limit, respectively. Stability analyses indicated the reactivity would be stable under perturbations of fuel, coolant, and inlet temperatures, with a positive reactivity insertion expected during reactor shutdown. This range corresponded to nuclear heating values of 150 to 650 MW and mass burn rates of 53 to 216 kg/year. The upper mass burn rate limit would require 1110 reactor years with a capacity factor of 0.9 to fission the global supply of MAs and could offset the annual U.S. MA production with eight reactors. Tritium breeding was assessed for keff = 0.9800 and 3.795% 6Li enrichment in the coolant, and a tritium breeding ratio of 1.602 0.017 was tallied, suggesting the reactor could, without elevated 6Li enrichment, produce tritium to both sustain operation and supply tritium for other fusion devices. Time-series modeling of fuel burnup was conducted for a four-batch loading scheme and three different fuel residence times at keff = 0.9800, which showed system performance would drop with burnup, and that the rate of this drop was lower for longer fuel residence times, motivating a means of reactivity control. Last, changes in fuel composition with burnup were assessed for relative concentrations of MAs, transmutation products, and fission products. The breeding of plutonium in the blanket was calculated and found to be of minimal proliferation concern.

Analysis of burnup performance and temperature coefficient for a small modular molten-salt reactor started with plutonium

Analysis of burnup performance and temperature coefficient for a small modular molten-salt reactor started with plutonium

Investigation on thermal conversion, kinetics and thermodynamics of coal gasification filter cake ash in oxygen fuel combustion

Coal gasification fine ash (CGFA) has the characteristics of high carbon content, fine-grained, and large annual output, causing environmental pollution and energy wastage. Oxygen fuel combustion is an effective way to solve the CGFA problem. In this paper, the effects of thermal conversion, kinetics and thermodynamics during the oxygen-fuel combustion of CGFA are analysed. The results show that the combustion of CGFA in an oxygen-fuel atmosphere significantly improves its burn-up performance and combustion stability, but the improvement in ignition performance is not significant. an increase in oxygen concentration of more than 30% in an O2/Ar atmosphere also improves the activation energy to some extent, but the improvement in combustion performance is not significant. The average activation energy of CGFA under 30% O2/70% Ar condition is 104.25 kJ/mol, and its combustion performance is comparable to that of bituminous coal blended gasification ash at 70%. A kinetic model was developed to describe the oxygen-fuel combustion process of FCA, which can be efficiently and cleanly burned under oxygen-fuel conditions and consumed on a largescale.

Core Study of 5-MW(thermal) Water-Cooled Super Small Modular Reactor (WSMR) Based on the Sensitivity Analysis of the Key Design Parameters for Operation

As a new type of nuclear power system, more attention has been paid to the small modular reactor in the development of nuclear power after the Fukushima accident. This paper focuses on the core study of a 5-MW(thermal) water-cooled super small modular reactor (WSMR) based on the sensitivity analysis of the key design parameters. The key design parameters include the fuel enrichment of 235U, the moderation performance, the separation work, and the layout scheme of the control rods. The impact analysis of the key design parameters on the steady-state performance and on the burnup performance is provided. From the impact analysis, the method of how to extend the operation life of the WSMR reactor is provided.

Investigation on Criticality and Burnup Performance of Pebble Bed Reactor with Thorium-based Nuclear Fuel

Thorium-based nuclear fuel has become an interesting subject for a variety of research with a wide range of applications. Research focusing on thorium-based fuel is aimed to overcome the scarcity and limitation of natural uranium resources as an alternative nuclear fuel in a thermal reactor. As thorium has no naturally occurring fissile isotope, it requires other fissile isotopes in order to be converted into fissile 233U to produce energy. The isotopes 235U and 2359Pu are two of the few alternatives available as the fissile nuclei for a thorium-fueled reactor. The purpose of this paper is to investigate the criticality and burnup performance of pebble bed reactor using two options of thorium-based fuel–namely, UO2-ThO2 and PuO2-ThO2. The HTR-10 was chosen as the reactor model. A series of criticality calculations with various uranium contents in UO2-ThO2 fuel and various plutonium contents in PuO2-ThO2 fuel was conducted using the Monte Carlo transport code MCNP6 and continuous energy nuclear data library ENDF/B-VII. The calculation result shows that 35% plutonium content in PuO2-ThO2 fuel has comparable criticality with 80% uranium content in UO2-ThO2 fuel. The former is shown to be better to ensure longer reactor operational time. However, that is due to the higher fissile fraction compared to that of the latter fuel. Meanwhile, thorium played little part in prolonging reactor criticality, as 233U production is not particularly significant.

Burnup Performance of CANDLE Burning Reactor Using Sodium Coolant

The CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) reactor concept was proposed to overcome the disadvantages of current reactor technologies. In this study, a Monte Carlo–based procedure is developed for quantitative comparison of burnup performance and neutronic characteristics between lead bismuth eutectic (LBE)–cooled and sodium-cooled CANDLE reactors to demonstrate the possibility of using sodium coolant in a small CANDLE burning reactor. In this procedure, a neutron transport equation is solved using the MVP code with the JENDL-4.0 library, and the burnup calculation is solved using the MVP-BURN code with the detailed burnup chain. To simulate the fuel-shuffling process, an auxiliary code was developed using Python. The results show that for the same fuel pin design and core volume, changing the coolant from LBE to sodium reduced the keff by 2.3% and the average discharge burnup by 15.6%, due to the softer neutron spectrum and larger neutron leakage fraction. It would be necessary to increase the fuel volume and core radius approximately 38% and 17%, respectively, for criticality in a sodium-cooled CANDLE core.

Computational study on distribution feasibility of radioactive waste transmutation in accelerator driven system

In this paper, the feasibility of high-level radioactive waste transmutation in accelerator driven system sub-critical reactor assembly, has been studied for two zone's model and with three different core configurations. The inner zone has a fast neutron spectrum and the outer one has a thermal neutron spectrum. The subcritical core is coupled with external neutron source of energy 14 MeV (D-T source). The effects of high level waste isotopes sample (238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm, and 245Cm) distribution on the neutron spectrum and burnup performance in the inner zone have been investigated and discussed, by proposed three core configurations non-uniform, uniform, and spiral. The burnup calculations have been performed for one-year operation cycle for all the all proposed models. This work shows that one can effectively transmute most of the actual minor actinides isotopes in the inner fast spectrum zone of the proposed system, with optimal distribution of these isotopes.

Preliminary investigation on the sodium fast reactor concave cores with near‐zero or negative sodium void reactivity

A preliminary investigation on the sodium fast reactor (SFR) concave cores was conducted with the aim of obtaining near-zero or negative sodium void reactivity (SVR). The geometry of the MOX fueled 750 MWe Japanese Sodium Fast Reactor (JSFR) core is transformed into concave cores by maintaining the same power level and heavy metal fuel inventory. The inner and outer core (with axial blanket in the lower part) configuration is maintained as in the original JSFR but the height of the inner core was set lower than the outer core to improve the SVR. Also a short inner blanket (20 cm thick) is inserted into the inner core for the same purpose. Radial blanket with the same height is also adopted as in the original JSFR to maintain the breeding ratio. In this investigation, the MOX fuel composition for the fresh fuel does not contain minor actinides. JOINT-FR code system with JENDL-4.0 library is used for neutronics, burnup, kinetics parameters as well as temperature and SVR calculations in 2-D R-Z geometry multigroup diffusion approximation. Several candidates of concave cores with favorable near-zero or negative SVR have been found whose burnup performance is comparable with the original JSFR core.

Evaluation on fuel cycle and loading scheme of the Indonesian experimental power reactor (RDE) design

Abstract In 2014, Indonesian government launched a program to build an Experimental Power Reactor, named Reaktor Daya Eksperimental (RDE), in South Tangerang, Banten. The RDE is a 10 MWt pebble-bed reactor utilizing TRISO fuel technology. A comprehensive neutronics and burnup performance evaluation on the RDE fuel loading schemes (once-through-then-out (OTTO) and multipass), fuel cycles (uranium and thorium) and optimum fuel compositions is conducted and reported in this paper. The optimum design parameters of the RDE are expected to contribute in minimizing the fuel cost, storage capacities, and other back-end environmental burden. The evaluation results show that OTTO fuel loading scheme is a better trade-off option considering the complexity of the mechanisms involved in the design. The optimal composition for uranium fuel cycle is 8 g heavy metal/pebble with U-235 content of 13.7 w/o for achieving the design target burnup of 80 GWd/t, while for thorium fuel cycle is 18.2 g/pebble with U-233 content of 7.2 w/o.

Simulated Microstructural and Compositional Evolution of U-Pu-Zr Alloys Using the Potts-Phase Field Modeling Technique

U-Pu-Zr alloys are considered ideal metallic fuels for experimental breeder reactors because of their superior material properties and potential for increased burnup performance. However, significant constituent redistribution has been observed in these alloys when irradiated, or subject to a thermal gradient, resulting in inhomogeneity of both composition and phase, which, in turn, alters the fuel performance. The hybrid Potts-phase field method is reformulated for ternary alloys in a thermal gradient and utilized to simulate and predict constituent redistribution and phase transformations in the U-Pu-Zr nuclear fuel system. Simulated evolution profiles for the U-16Pu-23Zr (at. pct) alloy show concentric zones that are compared with published experimental results; discrepancies in zone size are attributed to thermal profile differences and assumptions related to the diffusivity values used. Twenty-one alloys, over the entire ternary compositional spectrum, are also simulated to investigate the effects of alloy composition on constituent redistribution and phase transformations. The U-40Pu-20Zr (at. pct) alloy shows the most potential for compositional uniformity and phase homogeneity, throughout a thermal gradient, while remaining in the compositional range of feasible alloys.

Burnup performance of MOX and Pu-ROX fuels in small pebble bed reactor with accumulative fuel loading scheme

Pebble bed reactors (PBRs) are among the promising designs for future nuclear power plants. There are several fuel management strategies for PBRs. The accumulative fuel loading scheme was introduced as the simplest fueling strategies for a PBR. In this study, mixed-oxide (MOX) and plutonium-rock like oxide (Pu-ROX) fuels were introduced as starting materials. The MVP-BURN code was used to calculate the burnup performance of these fuels. The optimum fuel composition of MOX fuel was 5 g HM/pebble when uranium-plutonium ratio was 70:30. In the case of MOX fuel, the maximum discharged burnup was 172 GWd/t over 6 years of operation time. On the other hand, the optimum fuel composition for Pu-ROX fuel was 3 g HM/pebble. The results showed that the maximum discharged burnup was 550 GWd/t with an operation time of 12.7 years. The maximum excess reactivity occurred in the initial condition for both fuels. Moreover, the results of initial excess reactivity for both fuels were quite similar. The use of burnable poison materials to compensate for the initial excess reactivity of both fuels was not effective. A negative temperature coefficient was achieved for both MOX and Pu-ROX fuels throughout the operation period.

Study on Optimization Design for CSR1000 Core

An optimization design of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWe (CSR1000) conceptual core is proposed. Steady-state performance of the proposed core is then studied with the SCWR core steady-state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature, and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path, which is suited with assembly power, is also presented. The study shows that the refueling cycle of CSR1000 core can be extended effectively under the optimization design.

Burnup Performance of a PBR with an Accumulative Fuel Loading Scheme Utilizing Burnable Poison Particles in UO2 and ROX Fuels

The accumulative fuel loading scheme is a promising design for pebble bed reactors (PBRs). In a previous study, burnup analysis for such a reactor was performed using UO2 fuel and the results showed that the maximum burnup and operation period were 223 GWD/t and 10.2 years, respectively. In this study, rock-like oxide (ROX) fuel was analyzed in PBR with an accumulative fuel loading scheme. The results showed that a maximum burnup and operation period of 218 GWd/t and 8.4 years, respectively. However, high excess reactivity occurred in the initial condition for both UO2 and ROX fuels. In this study, burnable poison (BP) particles were used to suppress the initial excess reactivity. Several BP materials such as B4C, Gd2O3, Er2O3 and CdO were investigated. In case of UO2 fuel, the combination of B4C and Gd2O3 was found to be optimal for flattening the reactivity swing in the initial condition. However, due to the fissile density of ROX fuel being about five times lower than that of UO2 fuel, single-material BP particles (B4C) seemed to be optimal for flattening the reactivity swing. By utilizing the BP particles at the initial condition, the results of UO2 and ROX fuels showed that maximum keff was decreased from 1.423 to 1.068 and from 1.357 to 1.033, respectively. On the other hand, the maximum discharge burnup value and core lifetime were quite similar to the results for the reactor that did not use BP particles at the initial condition.

Impact of Neutron Spectrum Shift on Breed and Burn Reactor Concept

Neutronic analysis has been performed for an infinite fuel cell and a one-dimensional (1D) slab Breed and Burn (B&B) reactor with metallic fuel (U-10%wt Zr), cladding of HT-9, and coolant of lead bismuth eutectic (Pb-55.5%wt Bi). The purpose is to investigate the effect of the neutron spectrum on fuel burning in the B&B reactor and possibility to improve burnup performance of the B&B reactor by applying the neutron spectrum shift. The results of the fuel cell analysis indicate that inserting graphite moderator pins into a conventional fuel cell causes the neutron spectrum to shift into a softer one. This accelerates the transmutation of uranium-238 into plutonium-239 in the first burnup period. After a certain burnup period, the fuel cell can then provide a higher infinite neutron multiplication factor (k∞), on the condition that the moderator pins are withdrawn. We then present the neutronic analysis for a 1D slab B&B reactor. The primary results of the analysis indicate that it is possible to improve burnup performance by applying a neutron spectrum shift, and that the shift increased maximum burnup of the core by about 2.44%.

Alternative Fueling Scheme for the Indonesian Experimental Power Reactor (10 MWth Pebble-Bed HTGR)

The National Nuclear Energy Agency of Indonesia (BATAN) is launching a plan (2014) to build an 10 MWth Experimental Power Reactor (Reaktor Daya Eksperimental, RDE) in the Agency’s largest Research Center site, i.e. the Puspiptek Complex, Serpong, South Tangerang, Banten, as a first strategic milestone for the introduction of large scale nuclear power plant fleets into the country. The main objective of the plan is to demonstrate safe and reliable electricity and process heat generation from a nuclear reactor. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched uranium (LEU) UO2 TRISO fuel. Following our previous design study with multipass fueling scheme, a scoping study on fuel composition parameters, namely heavy metal (HM) loading per pebble and uranium enrichment was conducted for the simpler once-through-then-out (OTTO) fueling scheme. No need of burnup measurement devices and refueling mechanism to send back discharged fuel pebbles into the core, low power and power density, and simple operation of the RDE become strongest motivation for the present work. The objective function for the optimization is the fissile loading requirement per energy generated (kg/GWd). The optimal heavy metal loading was found around 8 g/pebble while the uranium enrichment corresponding to the 80 GWd/t discharge burnup constraint is approximately 13.7 w/o. Compared with the burnup calculations results of multipass fueling scheme, OTTO fueling scheme burnup performance is slightly inferior in that slightly higher enrichment and fissile loading per pebble are required to achieved the same average discharge burnup.

Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5g-HM/pebble with 20% 235U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218GWd/t, with a core lifetime of about 8.4years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO2 fuel. A negative temperature coefficient was achieved for both UO2 and ROX fuels throughout the operation period.

Burnup performance of small-sized long-life CANDLE high temperature gas-cooled reactors with U–Th–Pa fuel

The innovative CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy has been successfully applied to both fast and thermal reactors. In particular for thermal reactor applications, CANDLE block-type high temperature gas-cooled reactors (HTGRs) with either U–Pu or U–Th fuel cycle had been proposed and investigated for their simple and safe reactor operation, and the ease of designing a long life reactor. In this work, we investigated the use of Pa-231 mixed homogeneously in the (Th-232/U-233)O2 fuel kernel of the TRISO particles, that is the U–Th–Pa fuel, to obtain better CANDLE burnup performance. Some consideration on how to supply the required U-233 and Pa-231 was also thoroughly discussed. U–Th–Pa fuel decreases the required amount of natural Gd burnable poison in the fresh fuel for establishing a valid CANDLE HTGR design since Pa-231 has a large thermal absorption cross section. Besides the role as a burnable poison nuclide, Pa-231 also serves as a fertile nuclide during the CANDLE burning since Pa-231 is finally transmuted to a fissile U-233 nuclide. A promising analysis result shows that for U-233 and Pa-231 concentrations of 15 and 7.50w/o, respectively, the discharged burnup can be increased from 138 (U-Th fuel) to 149GWd/t. This extends the core life time about 16%, i.e. from 35 (U-Th fuel) to 41years with CANDLE active core height of 800cm and reactor thermal power of 30MWth.

Application of melt and refining procedures in the CANDLE reactor concept

The application of melt and refining procedures has demonstrated great potential to solve the fuel integrity problem in the high-burnup condition of CANDLE reactors. However, if the melt and refining procedures are applied during operation, the reactor might lose all the nuclide distribution in the fuel pins and CANDLE burning becomes impossible to achieve. In this study, the application of melt and refining was simulated in two cases to overcome the cladding limitation at ⩾200dpa. It became clear that if the number of axial regions for the melt and refining procedure is chosen properly, CANDLE burning is possible to achieve even if each region is homogenized by the procedure. In addition, the fission products released by the melt and refining procedure increase the burnup performance of the CANDLE core remarkably. It is also possible to improve the engineering design by reducing the fuel volume fraction to a minimum at 48%.